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  • 1. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES The quality of the instrumentation used in nuclear medicine is an important part of the optimization of medical exposure. According to the BSS the licensee shall ensure a proper image acquisition and processing (II.17). It is also stated that measurements of the imaging devices shall be done at the time of commissioning and periodically thereafter. This statement can be applied not only to the imaging devices but also to all types of equipment used in connection with nuclear medicine examinations and therapy. Records should be kept of the results of the quality control of all instruments used in the nuclear medicine department. Limits of acceptability for the different parameters tested should be established. 1. ACTIVITY METER AND CALIBRATION OF SOURCES According to the BSS the licensee shall ensure that (II.19): (a) the calibration of sources used for medical exposure be traceable to a Standards dosimetry laboratory (d) unsealed sources for nuclear medicine procedures be calibrated in terms of activity of the radiopharmaceutical to be administered, the activity being determined and recorded at the time of administration; It is obvious that in nuclear medicine needs a calibrated activity meteris needed. The calibration is generally made at the place of manufacture using standard sources with activities that are traceable to a standard laboratory. A certificate should accompany the instrument. The activity meter or dose calibrator is a well-type gas ionisation chamber. The activity is measured in terms of the ionisation current. The chamber is sealed under pressure and the gas is usually a noble gas with a high atomic number in order to increase the probability of photon interaction. In the associated electrometer the ionisation current is converted to a voltage signal, which is amplified, processed and displayed in units of activity. This is possible since for a given radionuclide, assuming a fixed geometry and a linear response, the ionisation current is directly proportional to the activity. However, the response of an ionisation chamber to the radiations from different radionuclides varies according to the types, energies and abundance of these radiations. Appropriate adjustment of the amplification of the voltage signal is thus necessary. This is normally done through a selector push-button. It is, of course, very important that the correct amplification for a certain radionuclide is selected. A wrong selection of radionuclide can result in an activity reading, which can be wrong by a factor 2-3. Quality control of the equipment should include: • Test of precision and accuracy • Test of linearity of activity response • Test of reproducibility • Check of background Background should be checked before each measurement. Reproducibility should be checked once a week, while the other tests can be performed at longer intervals (monthly, quarterly or yearly) depending on the condition of the instrument and previously recorded variations. The sources needed to perform the tests should include a sealed reference source certified to +/-5% overall uncertainty or less. Examples of suitable sources are Co57, Ba133, Cs137 and Co60. Also note the importance of having reproducible measurement geometry. 1
  • 2. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES Test of linearity of activity response is generally performed using a high activity Tc99m source and repeated measurements over several half-lives of the radionuclide. Another method includes the use of calibrated attenuators, which simulate different activities. 2. COUNTERS AND PROBES Counters (generally named beta-counters and gamma-counters) to measure samples of plasma or other body fluids are used in several laboratory tests such as determination of kidney function (GFR), Schilling test (vitamin B12), ferrokinetic studies, total body water, etc. Several radionuclides are used including tritium. The activity to be measured is generally very low, which means that the sensitivity of the instrument must be high and the background count-rate as low as possible. The instruments are often equipped with an automatic sample changer with a capacity of several hundred samples. Beta-counter In the beta-counter or liquid scintillation counter, the sample is dissolved in an organic scintillation solution. Due to the resulting 100% counting geometry and the absence of any detector window, this means that the counter has excellent properties in detecting radionuclides of low activity emitting low energy beta-particles, such as H3 and C14. The light photons from the sample are collected by two photomultipliers in coincidence. This arrangement will reduce the background due to thermal noise and only true scintillation events will be analysed and counted. The main problem in liquid scintillation counting is the varying counting efficiency due to quenching of scintillation events. This process is caused by chemical contamination of the sample and/or a coloured sample. This means that the counting efficiency has to be determined for every sample. Therefore a quality control of the instrument must include a control of the correction methods. Otherwise the QC methods will be the same as for any scintillation counter. The sources needed for QC of a liquid scintillation counter include calibrated sources of H3 and C14 with different counting efficiencies as well as a background sample. Gamma-counter The detector in a gamma-counter is a NaI(Tl)-crystal with either an axial well or a transverse hole, the sample to be measured being lowered into the well or positioned in the hole. This configuration will result in a good counting geometry. Gamma rays absorbed or scattered in the crystal cause light photons (scintillations) which, in turn, give rise to electrical pulses at the photocathode of the photomultiplier tube (PM-tube). These pulses are amplified in the PM-tube and are fed to the associated electronics for further amplification, pulse-height analysis and counting. One important feature of the scintillation detector is the proportionality between the height of the pulse from the PM-tube and the photon energy absorbed in the detector. The function of the pulse-height analyser is to accept pulses within a chosen range of pulse-heights (energy) for counting, rejecting all others. The acceptance range is usually determined by the settings of two discriminator controls defining its lower (W- ΔW/2) and upper level (W+ΔW/2), where ΔW is called the window width. Operating in this differential mode is particularly important in the measurement of a radionuclide emitting one predominant energy in order to reduce the background count rate. The use of more than one pulse-height analyser makes it possible to measure samples containing more than one radionuclide. 2
  • 3. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES Probes Single and multi-probe counting systems for gamma radiation measurements are mainly used for iodine uptake measurements and for renography. They are based on a solid cylindrical NaI(l) crystal provided with a lead collimator to confer the necessary directional characteristics. The detector assembly is usually mounted in an adjustable support allowing it to be positioned in relation to the patient. The electronic components in a probe system are the same as in the gamma-counter. A probe system used for uptake measurements needs to be calibrated according to generally accepted methods. Quality control of any scintillation detector system should include tests such as: • Energy calibration • Energy resolution and energy response • Sensitivity • Counting precision • Background count rate • Linearity of activity response 3. EQUIPMENT FOR MORPHOLOGICAL AND FUNCTIONAL STUDIES Rectilinear scanner In nuclear medicine, a rectilinear scanner is an instrument designed to produce a two-dimensional image of the distribution of a radiopharmaceutical in an organ. The region of interest is scanned in successive rectilinear passes using a shielded and collimated scintillation detector (NaI(Tl)), the scanner head. The collimator is generally of multi-hole focussing type. Associated electronics provide for amplification and pulse-height analysis as in any scintillation detector system. The output pulses from the pulse-height analyser are further transformed by a display processor and then fed to a display device rigidly connected to the scanner head so that it follows its movement. The device presents the image as a distribution of monochrome or coloured marks produced by an electromechanical tapper on paper or as shades of grey produced by a flashing light source on photographic film. Rectilinear scanners are complex electromechanical instruments with a number of partially independent components. The production of high-quality images requires that all be in good functioning order by setting up a programme for regular maintenance and quality control. Besides the general tests of a scintillation detector system, the quality control programme of the rectilinear scanner should also include: • Test of geometric resolution • Test of system linearity • Test of total performance • Check of imaging device Gamma camera The gamma camera is the most commonly used imaging device in nuclear medicine. It is used to examine the spatial and temporal distribution of a radiopharmaceutical in an organ. The distribution of gamma-rays from the organ is projected by a collimator onto the large scintillation detector, creating a pattern of scintillations that outlines the 3
  • 4. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES distribution of radioactivity in front of the collimator. A photomultiplier tube array viewing the back of the detector, and electronic positioning logic circuits determine the location of each scintillation event as it occurs in the detector. When an event falls within the selected energy window of the pulse-height analyser it is registered into the image matrix. Hence, the final image is built by single scintillation events. In order to reach an acceptable image quality the number of scintillation events should be in the order of 100,000 –1,000,000. A gamma camera is equipped with 1-3 detector heads. The acquisition of data is dependent on the type of examination performed. Static acquisition is the collection of a certain number of counts from an organ with the gamma camera in a fixed position. Dynamic acquisition is collection of data according to a certain time schedule. Whole body acquisition means that the camera or the patient is moving resulting in an image that covers the whole body. Tomographic acquisition can be done by making the camera rotate around the patient. After reconstruction tomographic slices in any angle can be presented. An option available in modern cameras is the possibility to run a two-headed camera in coincidence mode, which means that it can be used in positron emission tomography (PET). The quality of the gamma camera image depends on the performance of the instrument and the correct handling of the instrument. Factors affecting image quality and formation include parameters such as: • Distribution of the radiopharmaceutical • Collimator selection • Spatial resolution • Energy resolution and window setting • Uniformity • Spatial positioning at different energies • Scattered radiation • Attenuation • Noise An optimum use of the equipment requires a well-trained staff running it and a programme for regular maintenance and quality control. The quality control of the gamma camera should include tests of parameters such as: • Uniformity • Energy window setting and spectrum display • Energy resolution • Sensitivity • Pixel size • Center of rotation • Linearity • Geometric resolution • Count losses at high count-rates • Multiple window positioning • Background • Total performance Some of the parameters should be checked daily, such as the uniformity, background and energy window settings, while others should be checked at a lower frequency but never less than once a year. 4
  • 5. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES A set of phantoms and sources are necessary in order to perform the regular quality control. These include flood source, line source, bar phantom and a total performance phantom (planar or tomographic). 4. CLINICAL DOSIMETRY According to the BSS II.20: Registrants and licensees shall ensure that the following items be determined and documented: (d) In diagnosis or treatment with unsealed sources, representative absorbed doses to patients; Since, in nuclear medicine, the possibility to measure the absorbed dose in an organ is very limited, the dose estimates must rely on calculations. The Medical Internal Radiation Dose Committee (MIRD) of the U.S. Society of Nuclear Medicine has developed a scheme for the calculation of absorbed doses in nuclear medicine and this methodology has been widely adopted. It introduces the concept of source organ and target organ, the former being any organ in the body containing radioactivity and the latter being any organ which absorbs energy. Radiation emitted from within a source organ will always affect that same organ, so that one target to be considered will be the same as the source; in general, however, each source organ may affect several targets and vice versa. The MIRD scheme provides for the calculation, for any particular radionuclide, of the mean absorbed dose in any specified target organ, which results from a standard number of disintegrations in a specified source organ. This is known as the " mean dose per unit cumulated activity" or the "S-value". The cumulated activity in an organ is the total number of disintegrations that occur within the organ. Thus if an organ were to contain a constant activity, A o, for a fixed time, T (with nothing before or after this time) then the cumulated activity would be AoxT. A more realistic example is that of an organ containing an activity Ao at time 0 which then falls exponentially with half-life Teff. (This half-life Teff will usually result from both physical decay of the radionuclide and biological elimination.) In this case the cumulated activity will be 1.44xAoxTeff. For more complex time patterns of radionuclides within organs the biokinetic data must be acquired and analysed so that a cumulated activity figure for each organ can be determined either by summing exponential terms or by summing measured activity levels at suitable intervals. Multiplying the S-value by the cumulated activity ÃS in the source organ gives the mean absorbed dose in the target organ arising from one source. Summing the figures arising from all relevant source organs will give the mean absorbed dose in the target organ Dt. Dt = DsxÃsxSs,t where Ss,t represents the S-value for a particular pair of source and target organs, and the summation is made for all sources affecting the target. If both sides of the dose equation are divided by the total administered activity, it will be seen that the mean dose to the target organ per unit administered activity may be obtained by replacing the cumulated activity in the source organ by the residence time. The residence time is obtained by dividing the cumulated activity in an organ by the total activity administered to (or absorbed by) the patient or subject. If, 5
  • 6. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES following an injection of the radiopharmaceutical, a particular organ immediately takes up a fraction F of the administered activity, contains it for a time T and then releases it, then the residence time of the radionuclide in that organ will be FxT. In the more realistic example when the activity in the organ falls exponentially with half- life Teff, the residence time will be 1.44xFxTeff. Thus, the calculation of organ doses and hence the effective dose is separated into three distinct steps: 1) determine the cumulated activity (or residence time) for each organ (patient measurements, animal models, biokinetic models etc.); 2) calculate the S-value for each pair of source and target organs (given in the MIRD publications); and 3) use the dose equation to compute the mean dose to each target organ. 5. REFERENCES 1. INTERNATIONAL ATOMIC ENERGY AGENCY. Quality Control of Nuclear Medicine Instruments. IAEA-TECDOC-602, IAEA, Vienna (1991). 2. WORLD HEALTH ORGANIZATION and INTERNATIONAL ATOMIC ENERGY AGENCY. Manual on Radiation Protection in Hospital and General Practice. Vol. 4. Nuclear medicine (in press) 3. INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION Radiation dose to patients from radiopharmaceuticasl, ICRP Publication No. 53. Oxford, Pergamon Press, 1987 (Annals of the ICRP 18, 1-4). 4. INTERNATIONAL ELECTROTECHNICAL COMMISSION. Characteristics and test conditions of radionuclide imaging devices: Anger-type gamma cameras. IEC,Geneva, 1992 (Publication IEC 789). 5. AMERICAN ASSOCIATION OF PHYSICISTS IN MEDICINE. Scintillation camera acceptance testing and performance evaluation. AAPM, Chicago, 1980 (AAPM Report No. 6, Nuclear Medicine Committee). 6. AMERICAN ASSOCIATION OF PHYSICISTS IN MEDICINE Computer-aided scintillation camera acceptance testing. AAPM, Chicago, 1982, (AAPM Report No. 9). 7. INSTITUTE OF PHYSICAL SCIENCES IN MEDICINE Quality control of gamma cameras and associated computer systems. IPSM, York, 1992 (IPSM Report No. 66). 8. HOSPITAL PHYSICISTS ASSOCIATION. Quality control of nuclear medicine instrumentation, HPA, London,1983 ( HPA CRS-38). 9. HINE G.J., PARA P., WARR C.P. Measurements of the Performance Parameters of Gamma Cameras, Part I. Washington D. C., US Government Printing Office, 1977 (FDA Publication 78-8049). 10. HINE G.J. et al. Measurements of the Performance Parameters of Gamma Cameras, Part II. Washington D. C., US Government Printing Office, 1979 (FDA Publication 79-8049). 6
  • 7. IAEA RADIATION PROTECTION IN NUCLEAR MEDICINE PART 7. OPTIMIZATION OF MEDICAL EXPOSURE. DIAGNOSTIC PROCEDURES 11. KARP JS, MUEHLLNER G. Standards for performance measurements of PET scanners: Evaluation with the UGM PENN-PET 240H . Scanner Med. Prog. Technol 17: 173 - 187, (1991) 12. LOEVINGER R, BUDINGER TF WATSON EE. MIRD primer for absorbed dose calculations. New York, The Society of Nuclear Medicine, 1988. 13. PARAS P. Performance and quality control of nuclear medicine instrumentation. In: Proc. Int. Symp. Medical Radionuclide Imaging, Vol.II. Vienna, International Atomic Energy Agency, 1981. 14. TER-POGOSSIAN MM. Positron Emission Tomography (PET) in Principles of Nuclear medicine. Ed. H. Wagner, Z. Szabo, J W Buchanan. W.B. Saunders. (1995) p. 342-362 15. TODD-POKROPEK A. Quality Control, Detection and Display. In: Kuhl DE., editor, Principles of Radionuclide Emission Imaging. Oxford, Pergamon Press,1983. 7